ML17146B095

Rev 1 to Susquehanna Unit 2 Cycle 3 Reload Analysis Design & Safety Analyses.

ML17146B095
Person / Time
Site:	Susquehanna
Issue date:	11/30/1987
From:	Jason White ADVANCED MEDICAL SYSTEMS, INC.
To:
Shared Package
ML17146B090	List: ML17146B089 ML17146B092 ML17146B094 ML17146B095 ML17146B096 ML17146B102 ML18040B194
References
ANF-87-126, ANF-87-126-R01, ANF-87-126-R1, NUDOCS 8712310158
Download: ML17146B095 (47)

Text

ANF-87-1 26 REVIStON 1 AD~MHCSDo HUCIt.EARFUSM CORPORATION SUSQUEHANNA UNIT 2 CYCLE 3 RELOAD ANALYSIS DESIGN AND SAFETY ANALYSES.

NOVEMBER 1987 ANAFFII.IATEOF KRAFTWERK UNION Q~ KRU 87i2310i58 87i223 0500058]

ADOCK POR ~

ADVANCEDNUCLEARFUELS CORPORATION ANF-87-126 Revision 1 Issue Date: 11/25/87 SUSQUEHANNA UNIT 2 CYCLE 3 RELOAD ANALYSIS Design and Safety Analyses Prepared By:

J. A. White BWR Safety Analysis Licensing and Safety Engineering Fuel Engineering and Technical Services AIIAFFILIATEOF KRAFTWERK UNION Qxsvu

CUSTOMER DISCLAIMER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT PLEASE READ CAREFULLY Advanced Nuclear Fuels Corporation's warranties and representations con-ceming the subject matter of this document are those set forth in the Agreement between Advanced Nuclear Fuels Corporation and the Customer pursuant to which this document is issued. Accordingly, except as otherwise expressly pro-vided In such Agreement, neither Advanced Nuclear Fuels Corporation nor any person acting on its behalf makes any warranty or representation, expressed or implied, with respect to the accuracy, completeness, or usefulness of the infor-mation contained In this document, or that the use of any information, apparatus, method or process disclosed ln this document will not infringe privately owned rights: or assumes any liabilities with respect to the use of any information, ap-paratus, method or process disclosed in this document.

The information contained herein is for the sole use of Customer.

In order to avoid Impairment of rights of Advanced Nuclear Fuels Corporation in patents or inventions which may be included in the information contained in this document, the recipient, by its acceptance of this document, agrees not to publish or make public use (in the patent use of the term) of such information until so authorized in writing by Advanced Nuclear Fuels Corporation or until after six (6) months following termination or expiration of the aforesaid Agreement and any extension thereof, unless otherwise expressly provided in the Agreement. No rights or licenses In or to any patents are implied by the furnishing of this docu-ment.

XN NF F00.765 (1

ANF-87-126 Revision 1 TABLE OF CONTENTS Section Pacae

1.0 INTRODUCTION

. ~ ..............,....,................................ 1 2.0 FUEL MECHANICAL DESIGN ANALYSIS................................... 2 3.0 THERMAL HYDRAULIC DESIGN ANALYSIS.............. ~.................. 3 3.2 Hydraul i c Characteri zati on........................................ 3 3.2.1 Hydraul i c Compatibility........................................... 3 3.2.3 Fuel Centerline Temperature....................................... 3 3.2.5 Bypass Flowe ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ \~~~~~~~~~~~~~~~~~~~~~ ~ ~ ~ ~ ~ ~ ~ 3 3.3 MCPR Fuel Cladding Integrity Safety Limit........... ~ ~ ~ ~ ~ ~ ~ ~ ~ 0 ~ ~ ~ ~ 3 3.3.1 Coolant Thermodynamic Conditions 3 3.3.2 Design Basis Radial Power Distribution. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 4

.3.3 Design Basis Local Power Distribution. ~ ~ ~ ~ ~ ~ ~ \ ~ ~ ~ ~

4.0 NUCLEAR DESIGN ANALYSIS.. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 5 4.1 Fuel Bundle Nuclear Design Analysis....... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 5 4.2 Core Nuclear Design Analysis ~ ~ ~ ~ ~ ~ ~ 5 4.2.1 Core Configuration....... 5 4.2.2 Core Reactivity Characteristics...,....,.. 6 4.2.4 Core Hydrodynamic Stability..... ~ .

5.0 ANTICIPATED OPERATIONAL OCCURRENCES......, ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 7 5.1 Analysis Of Plant Transients At Rated Cond itlons ~ ~ ~ ~ ~ ~ ~ ~ ~ 7 5.2 Analyses For Reduced Flow Operation....... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 7 5.3 Analyses For Reduced Power Operation...... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 8 5.4 ASME Overpressurization Analysis.......... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 8 5.5 Control Rod Withdrawal Error (CRWE) 5.6 Fuel Loading Error........ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 9 5.7 Determination Of Thermal Margins.......... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 9 6.0 POSTULATED ACCIDENTS... 10 6.1 Loss-Of-Coolant Accident....,,...... 10 F 1.1 Break Location Spectrum........ 10

ANF-87-1 Revision TABLE OF CONTENTS (Continued)

Section Pacae 6.1.2 B reak Size Spectrum............................................... 10 6.1.3 H APLHGR Analyses.............................'..................... 10 6.2 Control Rod Drop Accident........,... 11 7.0 TECHNICAL SPECIFICATIONS........ 12 7.1 Limiting Safety System Settings...... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 12 T.1.1 HCPR Fuel Cladding Integrity Safety L imit 12 7.1.2 Steam Dome Pressure Safety Limit 12 7.2 Limiting Conditions For Operation. 12 7.2.1 Average Planar Linear Heat Generation Rate L'imits................. 12 7.2.2 Minimum Critical Power Ratio 1 7.2.3 HGR Llmlts ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ 1 73 Surveillance Requirements...... ..... 14 7.3.1 Scram Insertion Time Surveillance.... 14 7.3.2 Stability Surveillance.......... 14 8.0 METHODOLOGY REFERENCES.......... 15 9.0 ADDITIONAL REFERENCES....... 16 APPENDICES A. SINGLE LOOP OPERATION............. A-1 B. SEISMIC-LOCA EVALUATION....,.................,.........,.......... B-1

ANF-87-126 Revision 1 LIST OF TABLES Table Pacae

4. 1 Neutronic Design Values........................................... 23 B. 1 Comparison Of Physical And Structural Characteristics For 8x8 And 9x9 Fuel Assemblies......................... .. .. . B-2 LIST OF FIGURES Ficiur e Pacae 3.1 Susquehanna Unit 2 Cycle 3 Hydraulic Demand Curve Power vs. Flow.... 17 3.2 Susquehanna Unit 2 Cycle 3 Design Basis Radial Power.............. 18 3.3 Design Basis Local Power Distribution - ANF XN-2 9x9 Fuel......... 19 Design Basis Local Power Distribution - ANF XN-1 9x9 Fuel......... 20 3.5 Design Basis Local Power Distribution - GE 8x8R (Central)

Fuel. 21 3.6 Design Basis Local Power Distribution - GE (Peripheral) 8x8R Fuel.......... ~ ~ 0 ~ ~ ~ ~ ~ ~ ...... 22 4.1 Susquehanna Unit 2 Cycle 3 Enrichment Distribution For ANF92-344L-9G4 XN-2 Fuel Lattice. 24 4.2 Susquehanna Unit 2 Cycle 3 Enrichment Distribution For ANF92-344L-10G5 XN-2 Fuel Lattice. ~ ~ ~ ~ ~ ~ ~ ~ 25 4.3 Susquehanna Unit 2 Cycle 3 Reference Core Loading Plan... ~ ~ ~ ~ ~ ~ ~ ~ ~ 26 4.4 Susquehanna Unit 2 Cycle 3 - Core Power vs. Core Flow...... 27 5.1 Susquehanna Unit 2 Cycle 3 Control Rod Withdrawal Error Analysis Limiting Initial Control Rod Pattern.. 28 5.2 Susquehanna Unit 2 Cycle 3 Flow MCPR Operating Limit....... 29

~

fj

ANF-87-126 Revision 1

1.0 INTRODUCTION

This report provides the results of the analyses performed by Advanced Nuclear Fuels Corporation (ANF)* in support of the Cycle 3 reload for Susquehanna Unit 2, which is scheduled to commence operation in the spring of 1988. This report is intended to be used in conjunction with ANF topical report

~XN-Np- -191 A, 91 4, R 11 1, Nppti 1 1 1 1 N Company Methodology to BWR Reloads," which describes the analyses performed in support of this reload, identifies the methodology used for those analyses, and provides a generic reference list. However, LHGR mechanical design limits (Reference 9. 1) and plant transient simulation model developments (Reference 9.141 b 1 dbyANF b 4 t NRN P 1 F~F-Volume 4, Revi'sion 1. Both References 9. 1 and 9.2 have been approved by the NRC for use in referencing in license applications. Section numbers in this 9 t 1 9 dtd tt b 1 X-N- - fNJ, olume 4, Revision 1.

The Susquehanna Unit 2 Cycle 3 core will comprise a total of 764 fuel assemblies, including 236 unirradiated ANF XN-2 9x9 assemblies, 324 irradiated ANF XN-1 9x9 assemblies, 112 irradiated General Electric 8x8R fuel assemblies (central region), and 92 irradiated GE 8x8R assemblies in the peripheral region. The reference core configuration is described in Section 4.2.

The design and safety analyses reported in this document were based on the design and operational assumptions in effect for Susquehanna Unit 2 during the previous operating cycle. Additional information and the results of design studies covering the development of 9x9 fuel assemblies for BWR reloads are contained in Reference 9.3.

f ANF-87-126 Revision 1 2.0 FUEL MECHANICAL DESIGN ANALYSIS Applicable ANF Fuel Design Report: Reference 9. 1 To assure that the expected power history for the fuels to be irradiated during Cycle 3 of Susquehanna Unit 2 is bounded by the assumed power history in the fuel mechanical design analysis, LHGR operating limits (Figure 3.3 of Reference 9. 1) have been specified. In addition, an LHGR transient operating'imit for Anticipated Operating Occurrences (Figure 3.4 of Reference 9. 1) has been specified for ANF 9x9 fuel. Additional information on rod bow, as requested in the NRC's safety evaluation report for Reference 9. 1, has been transmitted in Reference 9.4.

ANF-87-126 Revision 1 3.0 THERMAL HYDRAULIC DESIGN ANALYSIS 3.2 H draul i c Char aeter i zat i on 3.2.1 H draulic Com atibilit Component hydraulic resistances for the constituent fuel types in the Susquehanna Unit 2 Cycle 3 core have been determined in single phase flow tests of full scale assemblies. Figure 3. 1 shows the hydraulic demand curves for ANF 9x9 fuel and GE 8x8R fuel in the Susquehanna Unit 2 core. The similar hydraulic performance indicates compatibility for co-residence in 'he Susquehanna Unit 2 core.

3.2 '

~ ~ Fuel Centerline Tem erature Applicable Generic Report Reference 9. 1

.2.2 ~21 Calculated Bypass Flow Fraction 10.1%

at 104% Power/100% Flow 3.3 MCPR Fuel Claddin Inte rit Safet Limit Safety Limit MCPR = 1.06 3.3.1 Coolant Thermod namic Condition Rated Thermal Power 3293 Mwt Feedwater Flowrate (at SLMCPR) 16. 1 Mlbm/hr Core Pressure (at SLMCPR) 1042.9 psia Feedwater Temperature 383'F

ANF-87-1 Revision 3.3.2 Desi n Basis Radial Power Distribution See Figure 3.2 3.3.3 Desi n Basis Local Power Distribution See Figures 3.3 through 3.6

ANF-87-126 Revision 1 4.0 NUCLEAR DESIGN ANALYSIS 4.1 Fue Bund e Nuclea Desi n Anal sis Assembly Average Enrichment 3. 33%

Radial Enrichment Distribution Figure 4. 1 and 4.2 Axial Enrichment Distribution Uniform 3.44%

with 6" natural uranium top blanket Burnable Poisons Figure 4. 1 and 4.2 Note: Burnabl e poi sons are distributed uniformly over the enriched length of the designated rods. The natural urania axial blanket sections do not contain burnable absorber material.

Non-Fueled Rods Figure 4.1 and 4.2 Neutronic Design Parameters Table 4. 1 4.2 Core Nuclear Desi n Anal sis';2.

1 Core Confi oration Figure 4.3 Core Exposure at EOC2, HWd/HTU 18350.7 Core Exposure at BOC3, MWd/HTU 10911.2 Core Exposure at EOC3, HWd/MTU 21740.8 Maximum Cycle 3 Licensing Exposure Limit, HWd/MTU 22076

ANF-87-12 Revision 4.2.2 ore Reactiv't Characteris ics BOC Cold K-effective, All Rods Out 1.11353 BOC Cold K-effective, Strongest Rod Out 0.98524 Reactivity Defect (R-Value) 0.00% rho Standby Liquid Control System Reactivity, Cold Conditions, 660 ppm 0.98348 I

4.2.4 Core H drod namic Stabilit Power/flow Map Figure 4.4 Power Flow State Points Deca Ratio COTRA 64/42* 0.82 69/47** 0.75 66/45** 0.75

• Two pump minimum flow - APRN Rod Block intercept point. Extended operation at lower flow is not allowed by Technical Specifications.

• • Operation at less than 45% flow requires APRH/LPRN surveillance. In addition, operation at power/flow, combinations above and to the left of the line connecting these two points requires APRH/LPRtl surveillance. See Figbre 4.4.

ANF-87-126 Revision 1 5.0 ANTICIPATED OPERATIONAL OCCURRENCES Applicable Generic Transient Analysis Methodology Report References 9.5 5 9.7 5.1 Anal sis Of Plant Transients At Rated Conditions Reference 9.6 Limiting Transient(s): Load Rejection Without Bypass (LRWB)

Feedwater Controller Failure (FWCF)

Loss of Feedwater Heating (LFWH)

% Rated  % Rated Maximum Maximum Maximum Pressure Del ta Event Power* Flow Heat Flux Power ,

I

~aia CPR** Model LRWB 100% 100% 116.2% 267% 1194 0.24 COTRANSA/

XCOBRA-T FWCF 100% 100% 1 16. 8% 233% 1179 0.23 COTRANSA/

XCOBRA-T LFWH 100% 100% 121 '% 123% 1078 0.16 PTSBWR3/

XCOBRA Single Loop Operation: Appendix A 5.2 Anal ses For Reduced Flow 0 eration Reference 9.6 Limiting Transient(s): Recirculation Flow Increase Transient (RFIT)

• 104% power used in analysis as design bases.

• • Delta-CPR results for most limiting fuel type.

ANF-87-1 Revision 5.3 Anal ses For Reduced Power 0 eration Reference 9.6 Limiting Transient(s): Feedwater Controller Failure (FWCF)

Delta CPR

% Power Transient ANF 9x9 GE 8x8R 104 FWCF 0.23 0.20 80 FWCF 0.25 0.23 65 FWCF 0.28 0.26 40 FWCF 0.31 0.28 5.4 ASME Over ressurization Anal sis Reference 9.6 Limiting Event Full MSIV Isolation Worst Single Failure Direct Sera Maximum Pressure 1297 psig Maximum Steam Dome Pressure 1281 psig 5.5 Control Rod Withdrawal Error CRWE Starting Control Rod Pattern for Analysis Figure 5.1 100% Flow Distance Withdrawn Delta Rod Block Settin ~ft CPR 105 4.0 0.22 106* 4.5 0.24 107 5.0 0.26 108* 5.0 0.26

• Rod Block Monitor settings recommended for Cycle 3 operation.

ANF-87-126 Revision 1 5.6 Fuel Loadin Error Maximum Delta CPR 0.16 5.7 Determination Of Thermal Har ins Summary of Thermal Margin Requirements Event Power Flow Delta CPR* MCPR Limit LRWB 1P0%** 100% 0.24 1.30 FWCF 1PP%** 100% 0.23 1.29 LFWH 1PP%9c* 100% 0.16 1.22 CRWE 100% 100% 0.24 at 106% RBH 1.30 0.26 at 108% RBM 1.32 HCPR Operating Limits at Rated Conditions MCPR 0 eratin Limit 1.30 at 106% RBM 1.32 at 108% RBM Reduced Flow MCPR Limits Figure 5.2 Power Dependent HCPR Operating Limit Results for Cycle 3:

Limiting Transient ANF 9x9 GE 8x8R 100*+/100 LRWB 1.30 1.27 80/100 FWCF 1.31 1.29 65/100 FWCF 1.34 1.32 40/100 FWCF 1.37 1.34

,i

• Delta CPR results for most limiting fuel type.

• • 104% power used in analysis as design bases.

10 ANF-87-126 Revi si on 1 6.0 POSTULATED ACCIDENTS 6.1 Loss-Of-Coolant Accident Sei smi c- LOCA: Appendix B

6. 1. 1 Break Location S ectrum Reference 9.8

6. 1.2 Break Size S ectrum Reference 9.8

6. 1.3 MAPLHGR Anal ses ANF 9x9 Fuel Reference 9.9 Limiting Break: Double-ended guillotine pipe break Recirculation pump discharge line 0.4 Discharge Coefficient

~F Bundle Average Peak Clad Peak Local Exposure MAPLHGR Temperature* MWR**

GWD MTU ~kw ft ~Percent 0 10.2 2060 3.9 5 10.2 2069 3.7 10 10.2 2121 3.7 15 10.2 2140 4.8 20 10.2 2147 5.2 25 9.6 2016 2.7 30 8.9 1839 1.0 35 8.2 1752 0.7 40 7.5 1676 0.5

• Peak clad temperatures for XN-1 and XN-2 fuel are bounded by these results.

• • Metal Water Reaction.

ll ANF-87-1 Revision 6.2 Control Rod Dro Accident Section 8.0 Dropped Control Rod Worth, mk 13.5 Doppler Coefficient, 1/k dk/dT -10.6 x(10) 6 Effective Delayed Neutron Fraction 0.0058 Four-Bundle Local Peaking Factor 1.34 maximum Deposited Fuel Rod Enthalpy, cal/gm 205 Number of Rods Exceeding 170 cal/gm (250

12 ANF-87-126 Revision 1 7.0 TECHNICAL SPECIFICATIONS 7.1 Limitin Safet S stem Settin s 7.1.1 MCPR Fuel Claddin Inte rit Safet Limit MCPR Safety Limit 1.06

7. 1.2 Steam Dome Pressure Safet Limit Pressure Safety Limit (as measured in steam dome) 1325 psig Analysis shows that a steam dome pressure safety limit of 1358 psig is allowed but the 1325 psig value used in Cycle 2 is to be conservatively retained.

7.2 Limitin Conditions For 0 er ation 7.2. 1 Avera e Planar Linear Heat Generation Rate Limits Bundle Average Exposure MAPLHGR Limits kw ft GWD MT ANF 9x9 Fuel 0 10.2 5 10.2 10 10.2 15 10.2 20 10.2 25 9.6 30 8.9 35 8.2 40 7.5

13 ANF-87-1 Revision 7.2.2 Minimum Critical Power Ratio MCPR Operating Limits at Rated Conditions:

MCPR 0 eratin Limit 1.30 at 106% RBM 1.32 at 108% RBM MCPR Operating Limits at Off-Rated Conditions:

At Reduced Flow Figure 5.2 Total Core Reduced Flow Recirculation Flow MCPR

% Rated 0 eratin Limit 100 1.12 96 1.14 92 1.16 83 1,20 76 1.23 60 1.31 50 1.44 40 1.61 At Reduced Power Reduced Power Power Level MCPR

% Rated 0 eratin Limit 100* 1.30 80 1.31 65 1.34 40 1.37

• 104% power used in analysis as design bases.

ANF-87-126 Revision I 7.2.3 LHGR Limits LHGR Limits Figures 3.3 and 3.4 of Reference 9. 1 7.3 Surveillance Re uirements 7.3.1 Scram Insertion Time Surveillance Thermal limits established in Section 5.0 are based on minimum acceptable scram insertion performance as defined in the Technical Specifications. No additional surveillance for scram insertion is required for validation of thermal limits.

.3.2

~ ~ Stabilit Surveillance Power/Flow Map Figure 4.4 The Unit 2 Cycle 2 Technical Specifications require APRM/LPRM surveillance to the left of the 45% Constant Flow line and above the 80% Rod Block line.

Based on core hydrodynamic stability analyses, operation at power/flow combinations above and to the left of the line connecting the 66% Power/45%

Flow and 69% Power/47% Flow points but below the APRM Rod Block line needs to be added to the APRM/LPRM surveillance requirement (see Section 4.2.4).

15 ANF-87-126 Revision 1 8.0 METHODOLOGY REFERENCES See XN-NF-80-19(P)(A), Volume 4, Revision 1 for complete bibliography.

\

q, "~r 0~ )

6 44

-I

16 ANF-87-126 Revision 1 9.0 ADDITIONAL REFERENCES

9. 1

~X---F,R.X,Addll "Generic Mechanical Design Washington, September 4, 1986.

for Exxon Nuclear 1

Jet FXCF Pump BWR l,lhhlFuel,"

Reload d,

9.2 "Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal 111<<N hd1 d P Revision 2, Advanced Nuclear Fuels Corporation, Richland, Washington, January, 1987.

9.3 "Demonstration of 9x9 Assemblies for BWRs," EPRI NP-3468, Electric Power Research Institute, Palo Alto, California, Hay 1, 1984.

9.4 Letter, G. N. Ward (ANF) to G. C. Lainas (NRC), "Additional Information on Rod Bow," serial no. GNW:021:87, dated March 11, 1987.

9.5

~h-p--,h "Exxon Nuclear Plant Transient 11 X,AddN1F1C Washington, November 16, 1981.

Methodology for Boiling Reactors,"

l,ltlhl Water d,

9.6

~ "Susquehanna Unit 2 Cycle 3 Plant Transient Analysis," ANF 125, Rev. 2,

~ Advanced Nuclear Fuels Corporation, Richland, Washington, November 1987.

i,"X~,

~

9.7 "XCOBRA-T:

A 2,

1 Advanced A Computer Nuclear Code Fuels for 1 >>d BWR Corporation, Transient Thermal-Hydraulic Core Richland, Washington, February 1987.

9 8

~ "Generic LOCA Break Spectrum Analysis BWR 3 8 4 with Hodified Low Pressure Coolant Injection Logic Using the EXEH Evaluation Model," XN-NF-

~84-117 P, Advanced Nuclear Fuels Corporation, Richland, Washington, December 1984.

9.9 "Susquehanna LOCA-ECCS Analysis HAPLHGR Results for ENC 9x9 Fuel," XN-NF-86-65, Advanced Nuclear Fuels Corporation, Richland, Washington, May 1986.

9. 10 "Principal Reload Fuel Design Parameters, Fuel Design, Susquehanna Unit 2 Reload XN-2," XN-NF-1058, Advanced Nuclear Fuels Corporation, Richland, Washington, March 1987, Formerly Exxon Nuclear Company.

Advanced Nuclear Fuels 9x9 0~)

General Electric O~ 8x8R I

~o p co o

C p

0 O

I

~oQ I

L p

0 q

>vCi I

.>o oo K

I 0

CO 100.00 . 105.00 TIO.OO 115.M 120.00 125.00 QO.M Q5.M 140.00 %5.00 150.00 Assembly Flow Rate, KLB/HR Figure 3. 1 Susquehanna Unit 2 Cycle 3 Hydraulic Oemand Curve Power vs. Flow

80 70 60 50 00 C)

CL So 20 10 0

0 0.2 0.0 0.6 0.8 1 1.2 RRDIFIL POHER PERKING Figure 3.2 Susquehanna 2 Cycle 3 Oesign Basis Radial Power

19 ANF-87-126 Revision 1

• ~

0.88  : 0.91  : 0.96  : 1.04  : 1.02  : 1.04  : 0.96  : 1.00 : 0.96  :

• ~

• ~

0.91  : 0.93  : 0.98 : 1.07  : 0.91  : 1.07  : 0.97  : 1.04 : 1.01

• ~

• ~

0.96  : 0.98  : 0.90 : 1.04  : 1.03  : 1.04  : 1.04  : 0.99 : 0.96  :

• ~

1.04 : 1.07  : 1.04 : 1.00  : 0.99 : 1,00  : 1.05 : 0.94 : 1.04

• ~

• : 1.02 : 0.91  : 1.03 0.99  : 0.00 : 0.98 : 1.05 : 1.07 : 1.04

• ~

• ~

1.04 : 1.07  : 1.04 : 1.00  : 0.98 : 0.00 : 1.03 : 0.94 : 1.05

• ~

0.96 : 0 '7  : 1 '4 : 1.05  : 1.05 : 1.03 : 1.06 : 1.00 : 0,97 1.00 : 1.04  : 0.99 : 0,94  : 1.07 : 0.94 : 1.00 : 0.94 1.01 0.96 : 1.01  : 0.96 : 1.04  : 1.04 : 1.05 : 0.97 : 1.01 0.97 Figure 3.3 Design Basis Local Power Distribution Advanced Nuclear Fuels XN-2 9X9 Fuel

20 ANF-87-1 Revisio

• ~

• ~

0.91  : 0.92  : 0.95 : 1.01  : 1.01 : 1.01 : 0.96 : 0.98  : 0.95

• ~

• ~

0.92  : 0.94  : 0.98 : 0.97  ; 1.05 : 0.95 : 0.99 : 0. 95 : 0.98

• ~

• ~

• ~

0.95  : 0.98  : 0,93 : 1.06  : 1.05 : 1.06 : 1.05 : 0. 97 : 0.96

• ~

• ~

1.01  : 0.97  : 1,06 : 1.03  : 1,03 : 1.04 : 1.07 : 1. 06 : 1.02

• ~

• ~

1.01  : F 05  : 1.05 : 1.03  : 0.00 : 1.01 : 1.07 : 1.06  : 1.01 1 01 ' 95 1.06 : 1.04  : 1.01 : 0 00 F  : 1.04 : 0.96  : 1.02 0.96  : 0.99 1.05 : 1.07  : 1.07 : 1.04 : 1.06 : 1. 00 : 0.96 0.98  : 0.95  : 0.97 : 1.06 : 1.06 : 0.96 1.00 : 0.95  : 0.98 0.95  : 0.98  : 0.96 : 1.02 : 1.01 : 1.02 : 0.96 : 0.98  : 0.96 Figure 3.4 Design Basis Local Power Distribution Advanced Nuclear Fuels XN-1 9X9 Fuel

21 ANF-87-126 Revision 1

• ~

1.03  : 1.00  : 1.00  : 1.00  : F 00 : 1.00  : 1.01  : 1.03

• ~

• ~

• : 1.00 : 0.98  : 1.00 1.02  : 1.02 : 1.03  : 1.00  : 1.01

• ~

• ~

• : 1.00 1.00  : 1.01 : 1.01  : 1.01 : 0.90 : 1.03  : 1.00  :

• ~

• ~

• : 1.00 : 1.02 1.01 : 0.89  : 0.00 : 1.01 : 1.02 : 1.00

• ~

• ~

• : 1.00 1.02 1.01 : 0.00  : 0.89 : 1.01 : 0.99 : 1.00

• ~

• ~

• : 1.00 1.03  : 0.90 1.01  : 1.01 : 0.98 : 1.00 : 1.00

• ~

1.01 : 1.00  : 1.03 : 1.02 0.99 : 1.00 : 0.98 : 1.00 1.03 : 1.01  : 1.00 : 1.00  : 1.00 1.00 : 1.00 : 1.03 Figure 3.5 Design Basis Local Power Distribution General Electric (Central) SXSR Fuel

22 ANF 'evisio

• ~

1.00 : 1,00  : 1.00  : 1.00  : 1.00 : 1.00  : 1.00 : 1.00 :

• ~

0

• ~

1.00 : 1.00  : 1.00  : 1.00  : 1.00 : 1.00  : 1.00 : 1.00 :

• ~

• ~

• 100 1.00  ; 1.00  : 1.00  : 1.00 : 1.00  : 1.00 : 1.00 :

• ~

• ~

• : 1.00 : 1.00 1.00  : 1.00  : 0.00 : 1.00  : 1.00 1.00 :

• ~

1.00 : 1.00  : 1.00  : 0.00  : 1.00 : 1.00  : 1.00 1.00

• 100 1.00  : 1.00  : 1.00 1.00 : 1.00  : 1.00 : 1.00

• ~

1.00 : 1.00  : 1.00  : 1.00  : 1.00 : 1.00  : 1.00 : 1.00 :

1.00 : 1.00  : 1.00  : 1.00  : 1.00 : 1.00  : 1.00 : 1.00 Figure 3.6 Design Basis Local Power Distribution General Electric (Peripheral) SXSR Fuel

23 ANF-87-126 Revision 1 TABLE 4. 1 NEUTRONIC DESIGN VALUES Fuel Pellet Reference 9.10 Fuel Rod Reference 9.10 Fuel Assembl Reference 9.10 Core Data Number of fuel assemblies '64 Rated thermal power, HW 3293 Rated core flow, Hlbm/hr 100 Core inlet subcooling, Btu/ibm 24.0 Hoderator temperature, F 548.8 Channel thickness, inch .080 Fuel assembly pitch, inch 6.00 Wide water gap thickness, inch 0.562 Narrow water gap thickness, inch 0.562 Control Rod Data Absorber material B4C Total blade span, inch 9.75 Total blade support span, inch 1.58 Blade thickness, inch 0.260 Blade face,-to-face internal dimension, inch 0.200 Absorber rods per blade 76 Absorber rod outside diameter, inch 0.188 Absorber rod inside diameter, inch 0.138 Absorber density, % of theoretical 70.0

24 ANF-87-126 Revision

• : LL  : L  : HL  : M  : N H  : HL  : HL HL M  : MH  : N* HH N* ' HL

~ 4

~

HL M ". H* H: H O': HH H: HL H: HH:

\

H: H: H H: H M H  : N*  : H  : H  : W  : HH  : H  : HH

'A' N: MH: H H NH W: MH H*

~

HL  : H*  : MH  : H H MH  : MH HL HL  : H  : H  : M~ MH  : H*  : M ML ML L  : HL  : HL  : H M: HL: HL-: L LL RODS ( 1) 1.45 W/0 U235 L RODS ( 5) 1.95 W/0 U235 HL RODS (16) 2.55 W/0 U235 H RODS (20) 3.27 W/0 U235 MH RODS (13) 4.23 W/0 U235 H RODS (15) 4.66 W/0 U235 H* RODS ( 9) 3.27 W/0 U235 + 4.00 W/0 GD203 W RODS ( 2) INERT WATER ROD Figure 4. 1 Susquehanna Unit 2 Cycle 3 Enrichment Distribution for the ANF92-344L-9G4 Xi4-2 Fuel Lattice

• • • • • • • • • • • • • 9: * **********

0 25 ANF-87-126 Revision 1

• ~

LL o L: \

ML M ML NL L

• ~

• ~

ML NH  : M*  : MH  : M* ML

• ~ N ML . N* M H: H MH M: NL

• ~

~ J o

~

MH H: H H- H: H N*: M N* ~

H W: "MH H: MH t*

~

~

M: NH: H H MH W  : NH  : M*  : M ML  : N~  : MH  : H  : H  : NH  : NH M: ML ML  : M  : N  : M*  : MH  : M*  : M ML ML :

ML  : NL  : M M: ML ML LL RODS ( 1) 1.45 W/0 U235 L RODS ( 5) 1.95 W/0 U235 ML RODS (16) 2.55 W/0 U235 M RODS (19) 3.27 W/0 U235 MH RODS (13) 4,23 W/0 U235 H RODS (15) 4.66 W/0 U235 M* RODS (10) 3.27'W/0 U235 + 5.00 W/0 GD203 W RODS ( 2) INERT WATER ROD igure 4.2 Susquehanna Unit 2 Cycle 3 Enrichment Distribution for the ANF92-344L-IOG5 XN-2 Fuel Lattice

26 ANF-87-12 Revi sio A2 C1 A2 C1 A2 C1 A2 C1 DO C1 A2 Ci EO C1 A2 C1 DO C1 DO C1 A2 C1 DO C1 A2 C1 FO C1 C1 A2 A2 C1 DO A2 DO Ci DO A2 00 C1 ~ DO C1 EO C1 A2 C1 00 A2 DO Ci 00 00 A2 00 C1 EO C1 C1 A2 A2 C1 DO C1 A2 C1 DO C1 DO A2 DO C1 EO C1 A2 C1 A2 C1 DO O'I C1 A2 DO A2 EO EO Ci C1 A2 A2 C1 00 C1 00 A2 C1 C1 00 C1 EO C1 EO C1 A2 C1 00 A2 DO Ci DO C1 EO C1 EO C1 00 C1 A2 DO C1 00 A2 00 A2 00 C1 EO C1 EO A2 82 C1 A2 DO A2 EO C1 EO C1 C1 C1 A2 A2 A2 Ci 00 DO EO C1 EO C1 A2 C1 EO 80 Ci EO C1 DO A2 A2 EO C1 EO Ci EO C1 EO C1 82 A2 C1 C1 C1 Ci C1 C1 C1 A2 A2 A2 A2 A2 A2 A2 A2 XY = Fuel Type X Burned Y Cycles

~Fuel T e Ho. of Buouieo Descri tion A 196 GE BX8 Type III 2.19 w/o U 235

~

8 8 GE BX8 Type II 1.76 M/o U.235 C' 324 XN. 1 ENC92.3318.7G4 140 XN.2 ANF92.333B.904 E 96 XN.2 ANF92.3338. 10G5 Figure 4.3 Susquehanna Unit 2 Cycle 3 Reference Core I.eading

27 ANF-87-126 Revision 1 120 110 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ h ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

100 APRN SCRAM re LIN) 90 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 4 ~ ~ ~ ~ ~ ~ ~

~ ~ ~ ~ ~ ~ ~

/e

~

(r ~ ~

APPM rr 100/v Xe ROD BLOCK: ~

/ R00.' IN 80 e r r ~

~ 4 ~ ~ ~ ~ ~ ~ ~ ~ ~ $ ~ ~ ~

~

ROO BiOCK N ~ / ~

MONITOR 70 ' 'r e

//

~ ~ ~ ~

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

) ~

~

e / 66/45) e 80 I ~

45K CORE PLOW I

W E 50 ~ ~

e e ~ ~ el' 80K R00 LINE 40 ~ ~ ~ 4' ~ ~ ~ ~ ~ ~ ~ ~

~ ~ 4 ~ ~ ~ ~ 4 ~

e 30 e ~ ~ ~ ~ e ~ ~

P ~ ~ ~ ~ ~ ~ 'I ~ ~ ~ ~ ~ ~ ~ ~ ~ 4 ~

20 ~ e ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ e NT CIRC '-PUMP M )N FL OW; 10 ~ ~ ~

0 0 10 20 30 40 50 60 70 80 90 100 CORE FLOW,  % RATED Figure 4.4 Susquehanna Unit. 2 Cycle 3 - Core Power vs. Core Flow

28 ANF 12

Revision 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 59 59 55 12 -- 00 -- 12 55 51 20 -- 26 -- 26 -- 20 51 47 00 -- 12 -- 08 -- 12 -- 00 47 43 -- -- 20 -- 20 20 20 43 39 -- 12 -- 08 -- 08 -,- 00 -- 08 08 -- 12 39 35 -- -- 26 44 44 26 35 31 -- 00 -- 04 -- 00 -- 00 -- 00 04 -- 00 31 27 -- -- 26 44 44 26 27 23 -- 12 -- 08 -- 08 -- 00* -- '8 08 -- 12 19 -- -- 20 -- 20 20 20 19 15 00 -- 12 -- 08 -- 12 -- 00 20 -- 26 -- 26 -- 20 12 -- 00 -- 12 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 Cycle Exposure 0.0 HHD/HTU Control Rod Density 23.3  %

Control Rod Being Withdrawn = 00*

Rod Fully Inserted =, 00 Rod Fully Withdrawn =--

Figure 5. 1 Susquehanna Unit 2 Cycle 3 Control Rod Withdrawal Error Analysis Limiting Initial Control Rod Pattern

1.60 Note: The MCPR operating limit shall be the maximum of 1.60 this curve, the full flow MCPR operating limit or the poorer dependent MCPR operating limit.

1.40 f4 1.30 O

A 1.80 O

A 1.10 40 50 60 70 80 90 100 TOTAL CORE RECIRCULATION FLOW (% RATED) figure 5.2 Susquehanna Unit 2 Cycle 3 Flow MCPR Operating Limit

Ah

+C"

A-1 ANF-87-126 Revision 1 APPENDIX A U

SINGLE LOOP OPERATION This Appendix provides limits and justification of those limits for Single Loop Operation (SLO).

A.l ANTICIPATED OPERATIONAL OCCURRENCES Reference A. 1 The NSSS supplier has provided analyses which demonstrate the safety of plant operation with a single recirculation loop out of service for an extended V

period of time. These analyses restrict the overall operation of the plant to lower bundle power levels and lower nodal power levels than are allowed. when oth recirculation systems are in oper ation. The physical interdependence between core power and recirculation flow rate inherently limits the core to less than rated power. ANF fuel was designed to be compatible with the co-resident fuel in thermal hydraulic, nuclear, and mechanical design performance. The ANF methodology has given results which are consistent with those of previous analyses for normal two-loop operation. Many analyses performed by the NSSS supplier for single loop operation are also applicable to single loop operation with fuel and analyses provided by ANF.

For single loop operation, the NSSS vendor found that an increase of 0.01 in the HCPR safety limit was needed to account for the increased flow measurement uncertainties and increased tip uncertainties associated with single pump operation. ANF has evaluated the effects of the increased flow measurement uncertainties on the safety limit HCPR and found that the NSSS vendor determined increase in the allowed safety limit MCPR is also applicable to ANF fuel during single loop operation. Thus, increasing the safety limit HCPR by 0.01 for single loop operation (1.07) with ANF fuel is sufficiently onservative to also bound the increased flow measurement uncertainties for single loop operation.

A-2 ANF Revisio The limiting MCPR operating limit for single loop operation is conservatively set using the limiting pump seizure accident delta CPR plus the single loop operation HCPR safety limit. This limit together with the. HCPRf curve for two loop operation plus .Ol and the MCPRp curve for two loop operation plus .Ol conservatively bound all transients.

The Technical Specifications require APRH/LPRH surveillance to the left of the 45% Constant Flow line and above the 80% Rod Block line. Based on core hydrodynamic stability analyses for Cycle 3, operation at power/flow combinations above and to the left of the line connecting the 66% Power/45%

Flow and 69% Power/47% Flow points needs to be added to the APRM/LPRM surveillance requirements. Figure 4.4 shows the core power versus core flow established for Cycle 3.

A-3 ANF-87-126 Revision 1 A.2 POSTULATED ACCIDENTS Reference A.2 ANF performed LOCA analyses for single loop conditions and has determined that the MAPLHGR limit curve (Section 7.2) for two-loop operation is also applicable to single loop operation for ANF 9x9 fuels.

A-4 ANF Revisio REFERENCES A. 1 "Susquehanna Unit 2 Cycle 2 Single Loop Operation Analysis," XN-NF 146, Advanced Nuclear Fuels Corporation; Richland, WA 99352, November 1986.

A.2 "Susquehanna LOCA Analysis for Single Loop Operation," XN-NF-86-125, Advanced Nuclear Fuels Corporation, Richland, WA 99352, November 1986.

B-1 ANF-87-126 Revision I APPENDIX B SEISMIC- LOCA EVALUATION The structural response of Advanced Nuclear Fuels Corporation's (ANF's) 9x9 fuel is similar to the structural response of the GE BxBR fuel it replaces in the Susquehanna Unit 2 core. Therefore, the seismic-LOCA structural response evaluation performed in support of the initial core remains applicable and continues to provide assurance that control blade insertion will not be inhibited following the occurrence of the design basis seismic-LOCA event.

The physical and structural properties of the 9x9 and the Bx8 fuel types which are important to the dynamic response of the fuel are summarized in Table B. l.

he close agreement between the important parameters for the ANF 9x9 and GE x8R fuel types indicates that the structural response would be very similar for both fuel types.

Similarity in the natural frequencies of the two fuel types mentioned above is further assured by the stiffness of the fuel assembly channel box. Both fuel types use the same fuel assembly channel box, and the channel box dominates the overall dynamic response of the incore fuel. ANF calculations show that approximately 97% of the stiffness of a fuel assembly is attributable to the stiffness of the channel box. For this reason, the dynamic structural response of the reload core is essentially that of the initial core, and the original seismic-LOCA analysis remains applicable. Deformation of the channel to the point that control blade insertion is inhibited is not predicted to occur.

B-2 ANF Revisio TABLE B. 1 COMPARISON OF PHYSICAL AND STRUCTURAL CHARACTERISTICS FOR 8X8 AND 9X9 FUEL ASSEMBLIES Fuel T es

~Pro ert ANF 9x9 GE 8x8R Assembly Weight, lbs 580 600 Number of Spacers Overall Assembly Length, in 171.29 171.40 Assembly Frequencies, cps Mode 1 1.9 2 3.7 3 6.5 10.4 5 15.5 6 21.9 7 29.1

• GE proprietary

ANF-87-126 Revision 1 Issue Date: 11/25/87 SUS(UEHANNA UNIT 2 CYCLE 3 RELOAD ANALYSIS Design and Safety Analyses Distribution:

D. A. Adkisson D. J. Braun R. E. Collingham L. J. Federico S. F. Gaines R. G. Grummer K. D. Hartley H. J. Hibbard S. E. Jensen T. H. Keheley J. N. Morgan L. A. Nielsen D. F. Richey G. L. Ritter C. J. Volmer J. AD White H. E. Williamson H. G. Shaw/PP8L (20)

Document Control (5)